The present invention relates to a fuel assembly adapted for a thermal neutron type reactor and more particularly to a fuel assembly utilizing fuel rods enriched in plutonium.
In view of effective usage of resource and energy security, there is a schedule for the utilization of plutonium recovered through reprocessing of spent fuel as a fuel in a thermal neutron reactor.
The plutonium radiates .alpha.-rays having high radiation intensity and it is hence necessary to prevent a human body from being internally exposed and also radiates neutrons and .gamma.-rays through decay and spontaneous fission. For this reason, production or fabrication of the fuel including the plutonium should be performed in a sealed environment in comparison with uranium fuel. In addition, many considerations must be paid for equipment and manufacturing processes. For example, multiple shielding equipment is required and strict attention should be paid for its decontamination and maintenance. Accordingly, it is extremely disadvantageous from economical and other view points to manufacture many kinds of fuel pellets and fuel rods containing the plutonium in different concentrations.
From another view point, since severe conditions are placed on the fuel rods containing the plutonium with respect to the conveyance, measurement control and criticality control, it is desired to reduce the number of fuel rods containing the plutonium by using a large containing ratio of plutonium in one fuel rod.
From the above view points, in a fuel assembly in which plutonium of a predetermined amount obtained through the reprocessing of spent fuel is utilized for a thermal neutron reactor, it is advantageous to substitute fuel rods each having high enrichment, in such a fuel assembly utilizing enriched uranium as shown in FIG. 7, with fuel rods containing the plutonium such as uranium-plutonium mixed-oxide fuel (MOX).
FIG. 7A shows a fuel arrangement in the radial direction and FIG. 7B shows a fuel arrangement in the axial direction. In these figures, reference numeral 1 denotes a channel box, 2 denotes a fuel rod, a symbol Ui (i=1-4) represents a uranium fuel, G is a fuel rod containing a burnable poison and W is a water rod.
In the substitution of the highly enriched uranium fuel rods in the uranium fuel rod assembly with the MOX fuel rods, the enrichment of the fissionable plutonium is experimentally set such that the reactivity characteristics as the plutonium fuel assembly and the peaking factors in the radial and axial directions becomes approximately the same as those of the uranium fuel assembly, but as a result, the following relationship will be established approximately.
Namely, in the case where the fuel rod having enrichment ei of the uranium fuel assembly is substituted with the MOX fuel rod having the enrichment Pi, supposing that the concentration of U-235 of the uranium of the MOX fuel is eB, the following equation (1) will be established. ##EQU1## I: Numbers of the fuel rods to be substituted.
In this equation Q becomes 1.2 to 1.5.
In another case where the fuel rod having high enrichment of the uranium fuel assembly is substituted with the MOX fuel rod, there is a case in which blanket portions having low concentration of U-235 are arranged to upper and lower ends of the uranium fuel rod. However, in the MOX fuel rods, there is a possibility of increasing the number of MOX fuel rods when a predetermined amount of the recovered plutonium is treated with the MOX fuel rod by providing the blanket portions having a lower U-235 concentration than natural uranium, thus being disadvantageous from an economical view.
In the thermal neutron reactor, when it is required to mix the plutonium with the uranium fuel and then to cause a fission reaction, nuclides other than U-235 for causing the fission reaction with respect to the thermal neutrons are plutonium isotopes of Pu-239 and Pu-241. The plutonium further includes Pu-240 absorbing the thermal neutrons and minute amount of Pu-238 and Pu-242.
The plutonium isotope Pu-241 is subjected to .beta.-decay with a relatively short half life (14.7 years) and decays to Am-241 as a neutron absorbing nuclide. During the cooling period of the spent fuel, Pu-238 and Pu-240 are transformed by .alpha.-decay of Cm-242 and Cm-244, respectively, but the transformed amounts thereof are small and their influence on the characteristics of the MOX fuel will be neglected.
The amount of fissionable isotope of plutonium to be enriched in the MOX fuel rod for a thermal neutron reactor is restricted to the amount necessary for keeping a chain reaction for a predetermined period and only the amount determined by design is mixed. Namely, U-235, Pu-239 and Pu-241 must be mixed so as to take a value designed for achieving a predetermined reactivity.
However, the plutonium is recovered by the reprocessing of the spent fuel and, accordingly, the isotope composition thereof differs in accordance with initial enrichment, burnup degree and cooling period of the spent fuel. FIG. 8 represents one example of the change of plutonium isotope in case of uranium fuel burnup.
Accordingly, the plutonium recovered after reprocessing through mixing of various kinds of spent O.sub.2 fuels is considered to have a certain isotopic composition due to reprocessing.
In general, the fuel design is determined so that the enrichment of the fissionable substance of the fuel rod has a predetermined reactivity before the recovery of the plutonium. However, it is complicated and disadvantageous to redesign the enrichment of the fissionable substance every time when the composition of the fissionable nuclide in the actually recovered plutonium differs in order to maintain the predetermined reactivity.
Further, since Am-241 is removed as an impurity at the time of the recovery of the plutonium during the reprocessing of the spent fuel, it is not necessary to consider accumulation thereof during the spent fuel cooling period. However, the Am-241 is accumulated in a period from the recovery of the plutonium to its loading into a core as a manufactured MOX fuel, so that it becomes necessary to design the enrichment in consideration of its influence on the core characteristics.
For the design of the enrichment of the MOX fuel, the use of the mixture of the recovered plutonium is assumed as described above, but the actual mixing ratio and the amount of Pu-241 to be transformed to Am-241 through the .beta.-decay during the cooling period and a time before its loading into the core after the reprocessing, are not clear. For this reason, in the conventional technology, the design has been made by tentatively assuming the containing rate of the sum Puf of the fissionable plutonium isotope Pu-239 and Pu-241 contained in the recovered plutonium by considering the initial enrichment, burnup degree, cooling period, and reprocessing amount, and the time after reprocessing before loading into the core as a fuel assembly after the transformation to PuO.sub.2 and manufacturing as the MOX fuel through mixing with UO.sub.2.
That is, at a time when the MOX fuel using the recovered plutonium after the reprocessing of the spent fuel in a boiling water reactor BWR, on the assumption that plutonium would be obtained having a plutonium containing ratio F of from about 80%, which is recovered after reprocessing spent fuel having a low initial enrichment and low burnup degree such as initially loaded fuel, to about 60%, which is recovered after the reprocessing of spent fuel having a high initial enrichment and high burnup degree such as reloaded fuel; the larger containment rate of about 80% is sought. This is because the reactivity becomes high in the case of the high containing rate F, that is, less amount of Pu-240 or Pu-242 as the thermal neutron absorbing nuclide, and accordingly, the margin with respect to the thermally limited value in the operational characteristic at loading in the core of the MOX fuel becomes small. According to the confirmation of the margin in the design on the assumption of such plutonium containing rate F, in the actual plutonium usage, in case the plutonium containing rate F could be made to be larger than the assumed value in the design, the margin with respect to the thermally limited value could be made large.
Further, with respect to the decay of Pu-241 and its daughter of Am-241 which is accumulated during the time from manufacturing the recovered plutonium into the MOX fuel and loading into the core, a shorter arrangement such as about one year has been made in consideration of an actually usable period. This is also based on the consideration that Pu-241 is contained in a higher amount in the less usable period and a higher margin can be ensured in view of its reactivity.
As described above, in the design of the MOX fuel, the enrichment should be set by preliminary assumption of the containing rate F with a certain composition in the initial design in spite of the fact that the composition of the plutonium and, in particular, the containing rate F of fissionable plutonium, to be actually obtained and used are not clear, and accordingly, in the initial design, the containing rate F should be set to a considerably large value with respect to the plutonium to be actually used.
After the MOX fuel has been designed and the margin having an optimum operational characteristic has been confirmed, the plutonium to be actually used is obtained. In a case where the containing rate F of the plutonium obtained is different from that of the assumption at the design time, the mixing amount of the plutonium is regulated so that the amount of fissionable nuclides .sup.235 U+.sup.239 Pu+.sup.241 Pu are to coincide. In this method, however, the reactivity is increased or decreased largely in accordance with the increasing or decreasing of the amounts of Pu-240 and Pu-242. In the case of the amount of PU-240 and Pu-242 being larger than that in the design, less reaction may be caused and power may be hence reduced. From the viewpoint of ensuring the design margin, since safety is maintained in the case of less reaction than in the case of excessive reaction, the design is made as a countermeasure so that such isotopes as Pu-240 have the containing rate less than the assumed plutonium composition and, in the case where the reactivity shortage is actually caused, an amount of fuel to be exchanged is increased or the running period is made short. However, such countermeasures result in the change of the fuel amount to be required or disadvantages in economy.
In the BWR as one example of the thermal neutron reactor, as shown in FIG. 9, a number of fuel assemblies are allocated within a channel box to constitute a core. Referring to FIG. 9, reference numerals 3, 4, 5, 6 and 7 denote a fuel assembly, a local power range monitoring system LPRM, an intermediate power range monitoring system IRM, a source range monitoring system SRM and a control rod, respectively.
In a space between the adjacent fuel assemblies 3, there is located a water gap area, having a constant width, for arranging a cross-shaped control rod, i.e. control blade, or instrumentation tube.
Coolant in the channel box constitutes two layer flows of water and steam during the running of the core, but in the water gap area, the coolant is not directly heated by the fuel rod, thus not generating steam. For this reason, the atomic density of hydrogen in the water gap area is large and the radical distribution of the thermal neutrons in the horizontal cross-section of the fuel assembly 3 of the BWR is made large in the peripheral portion of the fuel assembly. In order to make a power peaking factor small in an inner radial direction of the fuel assembly, it is adapted to arrange fuel rods having low enrichment to the peripheral portions of the fuel assembly as shown in FIGS. 7A and 7B.
In a case where the fuel rods containing plutonium are used, in order to make a power peaking factor small in the radial and axial directions of the fuel assembly, it is necessary to adjust the enrichment and the distribution of the uranium fuel rods and the density and the distribution of a burnable poison as well as their arrangement in the fuel assembly.
In general, the isotope of the plutonium recovered by the reprocessing is different, as described hereinbefore, in initial enrichment, burnup hysteresis, burnup degree, cooling period, etc. of the spent fuel reprocessed. However, when MOX fuel manufactured by mixing such plutonium with uranium is irradiated by thermal neutrons in the thermal neutron reactor, the isotopes of uranium and plutonium are transformed as follows. Namely, neutrons of the uranium fuel are absorbed by U-238 in the thermal neutron reactor and transformed into Pu-239. The Pu-239 is fissioned by the absorption of thermal neutrons, but a portion thereof is transformed into Pu-240, which is then transformed into Pu-241 by the absorption of neutrons. The Pu-241 is a fissionable nuclide, but a portion thereof is further transformed into Pu-242 by further absorption of neutrons. The Pu-241 is then transformed into Cm-242 by the absorption of neutrons. As a consequence of such reactions of the plutonium isotopes, the fissionable substance contained in the fuel rod is less reduced by the fission and the lowering of the reactivity does not progress due to the burnup of the fuel in comparison with the uranium fuel. Therefore, the power peaking of the fuel rod containing the plutonium in the radial direction of the fuel assembly has a tendency of being made large during the burnup in comparison with the fuel rod containing the uranium.
Further, since the composition of the plutonium isotope recovered after the reprocessing cannot be specified, it is necessary to either not utilize the rod in a case where the composition of the plutonium actually obtained includes a larger amount than the Puf amount assumed in the design, or to mix the recovered plutonium with other plutonium containing a lesser amount of Puf. In a conventional technique, there is no clear standard for judgement with respect to a mixing ratio and compositions in such mixing of the plutonium isotopes, and accordingly, such mixing has been performed case by case on the basis of experimental results.
As described above, in the conventional technology, when plutonium is mixed, there is no clear standard for judgement as to the mixing ratio and the composition to be obtained, so that the reactivity of the MOX fuel manufactured using the plutonium depends largely on the composition of the obtained plutonium, and the scattering of the reactivity is observed from a view point of interchangeability with the uranium, thus including disadvantageous points relating to the limit of the power peaking in core operation and in fuel economy.
The present invention was conceived in view of the above defects and disadvantages and aims to provide a fuel assembly for a thermal neutron type reactor capable of reducing the kind and number of fuel rods containing plutonium and increasing the margin with respect to the thermal limit during the running of the reactor.
Another object of the present invention is to provide a fuel assembly for a thermal neutron type reactor capable of easily setting the fission reaction effects of the fuel containing the plutonium and easily performing correction for maintaining its characteristics.
A further object of the present invention is to provide a fuel assembly for a thermal neutron reactor capable of preventing an excessive increase of power peaking factor in the radial and axial directions of the fuel assembly containing the plutonium.